During severe reactor accident scenarios, boron carbide (B4C), widely used as a neutron absorber in control rods, can undergo eutectic interactions with stainless steel (SS), which serves as cladding or structural material. This eutectic reaction induces low-temperature melting well below the intrinsic melting points of both components, potentially leading to premature control rod failure. Accurate characterization of this phenomenon is therefore critical for modeling core degradation during severe accidents. In this study, the eutectic melting behavior of the B4C-316L SS system under rapid heating was systematically investigated using a self-designed high-temperature apparatus combined with post-mortem characterization. Scanning electron microscopy (SEM), energy-dispersive spectroscopy (EDS), X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), Electron back scatter diffraction (EBSD) and CALPHAD-based phase diagram calculations were employed to analyze microstructures, elemental diffusion, and transformation pathways. The results demonstrate that rapid heating leads to transient diffusion at the reaction interface, shifting the effective eutectic temperature from 1203 degrees C at 80 degrees C/min to 1230 degrees C at 110 degrees C /min. Correspondingly, the diffusion layer thickness decreases from 1.64 to 1.13 mm. The observed phases, including Fe2B, Cr2B, and CrB, are consistent with CALPHAD predictions. This study elucidates the eutectic melting characteristics of the B4C-SS system under high heating rates, providing important micro-structural and thermodynamic insights for severe accident modeling.

Aerosol particles are critical pollutants that threaten environmental quality and human health. In severe accident conditions of nuclear power plants, containment spray systems are employed to mitigate fission product aerosols. The mechanical removal efficiency of aerosols is related to the parameters of droplets and particles. To quantify mechanical scavenging by hollow-cone sprays, systematic experiments were performed under 0.4 to 0.8 m3/h flow rates and 1 to 5 bar pressures. The results showed two characteristic regimes in the size-resolved removal efficiency: a small-particle plateau and a power-law increase for larger particles. Higher spray flow rates, even with identical total water volume, generate smaller droplets, higher droplet number density, and greater initial kinetic energy, thereby enhancing inertial impaction, interception, and wake-vortex capture. Evaluation of commonly used Slinn and Powers model families reveals consistent underprediction in both the plateau and inertial-dominated regions. Improved semi-empirical formulations for inertial impaction and wake-vortex capture were developed and coupled with established interception and diffusion expressions into a unified single-droplet framework. The refined composite model reproduces global and size-resolved decay across spray conditions and aligns with TOSQAN AG11/AG12 benchmarks, with overall prediction errors within +/- 50%. These findings provide a validated physical basis for modeling spray-driven aerosol removal in accidental aerosol release and containment safety applications.

The rupture of some high-energy coolant pipes in the reactor will cause high-temperature and high-pressure coolant to discharge into the low-pressure area, forming an under-expanded flashing jet. The substantial steam generated by flash evaporation can cause a rapid increase in pressure and may lead to local overpressure. In this paper, a virtual source method to simulate the highly under-expanded flashing jet in a confined vacuum condition is established, considering the effect of the relaxation of flash evaporation and rapid change in back pressure within the confined space on the flash fraction. A three-dimensional analysis model of the Ingress of Coolant Event experiment is established, and the virtual source method is verified with peak pressure and its arrival time. Meanwhile, the three-dimensional spatial distribution of pressure and temperature is achieved. The results indicate that the predictions of peak pressure, peak arrival time and the peak temperature during the discharging process are consistent with the experimental results. Due to jet impingement, a local high-pressure area appears near the vessel wall. The fluid temperature is lower than in other regions due to expansion and flash evaporation. This inhomogeneity gradually weaken as the fluid reaches saturated condition. The proposed virtual source method solves the problem of significant deviation in predicting flash fraction under the scenario of rapid back pressure change in the virtual source theory based on the constant back pressure assumption.

This paper developed a simplified analysis model that includes a reactor primary loop, a secondary loop, and a passive residual heat removal system. Simulations were conducted to evaluate the startup of the passive residual heat removal system under two different hot standby conditions. When all of the pipes are filled with low-temperature, low-pressure water in the hot standby condition, the flow rate takes around 100 s to stabilize after start. In another hot standby condition, the pipes connected to the main steam line are filled with nitrogen gas, and a venting operation is required before startup. However, after the venting is completed, there are no significant fluctuations in the flow rate. Although the system in both hot standby states can meet the long-term heat removal requirements, considering the advantages of simple maintenance of water-filled pipelines, a parameter sensitivity study was conducted for the hot standby state with water, to investigate the effect of isolation valve operation time, initial liquid column temperature, and loop resistance on the transient response of the system. Longer valve opening times, higher internal friction, and lower initial liquid column temperatures extend the time for flow stabilization and delay the establishment of a stable flow state.

Accurate calculation of steam condensation heat transfer coefficient (HTC) is critical for the design of large pressurized water reactors and the analysis of accident progression. To improve prediction accuracy and model robustness under complex non-condensable gas (NCG) conditions, this study develops a residual fully connected neural network (Res-FCNN) with a conventional fully connected neural network (FCNN) for comparison. The models predict HTC using six input parameters and are trained and evaluated on a database of 1013 data points from vertical tube steam condensation experiments. In a comprehensive comparison, the Res-FCNN clearly outperforms the conventional FCNN, achieving an average MAPE of 3.88% and 92.21% of predictions within +/- 10% deviation across 1000 independent trials. Independent validation with extrapolation is also conducted, demonstrating the greater superiority of Res-FCNN. Interpretability analyses using SHAP, PDP and ICE plots identifies the NCG mass fraction as the most dominant factor, and the feature-target relationships are consistent with prior research results, confirming the physical consistency and sound decision logic of the Res-FCNN model. The integration of deep learning with interpretability enhances model reliability and practical applicability, offering a worthwhile approach for advancing nuclear thermal analysis.

Eutectic melting behavior between nuclear materials is a critical issue in severe accidents. The eutectic reaction between boron carbide (B₄C) and stainless steel (SS) can lead to the early melting of control rods, forming a eutectic melt that may spread widely. The eutectic melting temperature varies with heating rate, especially under severe accident conditions, where control rod heated rapidly. In this study, differential thermal analysis and Thermogravimetric analysis were used to investigate eutectic melting behavior. Bulk B₄C and 316L SS samples were heated at rates of 2.5–20 K/min up to 1673 K. The eutectic melting temperature at different heating rates were obtained by the DTA curves and TG curves of the samples, revealing a trend of increasing eutectic melting temperature with higher heating rates. Post-experimental SEM and EDS analyses offered comprehensive insights into the microstructure and elemental distribution at the solidification interface. Results show a distinct three-layer structure across the cross-section of the melt, with a central eutectic diffusion layer. At lower heating rates, a dense and uniform diffusion layer was observed, whereas higher rates resulted in a rougher, more dispersed diffusion layer. The thickness of the diffusion layer decreased as heating rate increased. And molar concentration analysis of elements within the diffusion layer indicated a reduction in Fe molar percentage in the B-C-Fe-Cr quaternary system. This study provides critical support for understanding eutectic melting behavior and modeling eutectic reactions under severe nuclear reactor accident conditions.

In response to the design requirements of the safety discharge system for the primary loop of NHR200-II small onshore reactor, an experimental simulation device for the transient discharge system was established, with full pressure and flow rate during the air clearing process. Experimental study on the transient dynamic loading behavior of the water tank were conducted under the initial discharge pressure of 10 MPa. The influences of the submergence of sparger, pool temperature, discharge gas temperature, non-condensable gas fraction, and valve opening time on the amplitude and main frequency of dynamic load oscillation were obtained. The dynamic pressure amplitude under all conditions is between 23.8 kPa and 48.7 kPa, with a main frequency between 2.3 Hz to 3.1 Hz, showing low frequency and high amplitude characteristics. The findings indicated that the oscillation characteristics were most significantly influenced by submergence depth of the bubbler and the duration of valve opening, followed by water temperature, non-condensable gas fraction and discharge gas temperature.

The aerosol pool scrubbing efficiency has various applications in nuclear reactor systems, including the secondary side of steam generators, suppression pools, and spent fuel pools, discharging gas mixtures containing aerosols into a pool and allowing radioactive substances to remain in the pool, thereby reducing the discharge of radioactive materials into surroundings. The effectiveness of aerosol pool scrubbing depends on the gas–liquid dynamics observed in injection and rising zones. This paper presents an experimental investigation into the effect of submergence depth on effectiveness of aerosol pool scrubbing under bubble flow and jet flow. The efficiency of aerosol removal increases with increasing submergence depth for both flow patterns. Considering the behavior of bubbles in rising zone and key deposition mechanisms of centrifugation, bubble breakup, an analytical method for the aerosol scrubbing removal efficiency is developed combining the existing aerosol removal models under bubble and jet flow patterns. The model predictions are compared with experimental data and existing empirical formulas. According to the calculation results, it is found that the dominant mechanism of rising zone is bubble breakup and centrifugal deposition, and the reason for lower Decontamination Factor (DF) in rising region is due to low predicted value of bubble rising velocity model.

Aerosol retention inside narrow channels is the optimization direction of the leakage source term assessment for nuclear power plant containment. Based on the flow characteristics of carrier gas and the deposition characteristics of transported aerosol, a one-dimensional analysis method of aerosol retention in narrow channels is developed through considering different deposition mechanisms of inlet loss, gravity settlement, Brownian diffusion, turbulent deposition and steam condensation. The flow models of carrier gas and the retention models of aerosol are analyzed and verified, respectively. The flow of carrier gas deviates from laminar flow earlier through using the drag model of narrow channels. The prediction accuracy of aerosol penetration factor calculated by current analysis method in narrow channels is improved under laminar flow and turbulent flow through comparing with the previous calculation methods. Aerosol retention analysis is conducted on the narrow channels of steel containment under the typical severe accident. The turbulent deposition introduced by larger leakage channels increases the aerosols retention effect in narrow channels.

During the progression of a hypothetical core melt accident in light-water reactors, a porous debris bed forms in the lower head of the reactor vessel due to thermal shock or fuel coolant interaction. The two-phase flow pressure drop is essential to the cooling limitation of the debris bed. A consolidated database of two-phase flow through debris beds was established, and five typical two-phase resistance models were evaluated. The results indicate that the TS model's prediction error, accounting for the flow patterns, is smallest in the low void fraction region, with a mean absolute error of 3.81%. In the high void fraction region, current models fail to capture the behavior of the interfacial drag increasing again. The accuracy of interfacial drag prediction in high void fraction impacts the evaluation of the debris bed dryout process and determines the air-water co-current inflow rate under the mitigation measures. Introducing boundaries between the channel and annular flow patterns and modifying the two-phase flow resistance in the high void fraction based on annular flow characteristics in porous media establishes a modified model for low and high void fraction regions. When compared to 703 experimental data points, 97.16% of the experimental data falls within the 10% error range, with a mean absolute error of 4.06%.

In-vessel Loss of Coolant Accident (LOCA) is one of the fundamental design-basis accidents that must be considered in the design phase of the China Fusion Engineering Test Reactor (CFETR). If high-temperature coolant leaks into the Vacuum Vessel (VV), a flashing jet is formed, leading to localized pressure and temperature peaks. As the VV serves as a primary barrier for radioactive protection, these peaks may pose significant challenges to its integrity. In this study, a three-dimensional simulation model of the CFETR VV is developed to analyze the evolution of flow, pressure, and temperature fields during an in-vessel LOCA. The study investigates the structural evolution of the highly under-expanded jet and its impact on pressure and temperature dynamics. The results indicate that the pressure at the jet impact surface is significantly higher than at other locations within the VV. Comparisons of temperature and pressure variations at different monitoring points reveal that if the bursting valve is positioned at the furthest location, the pressure at the jet impingement surface may reach the VV’s pressure limit before the VV Pressure Suppression System is activated, as determined using the lumped-parameter analysis method. Additionally, temperature rises caused by gas compression at wave crossings pose challenges to the VV material’s thermal limits. Furthermore, a simulation with a larger breakage size of the First Wall coolant pipes is conducted. The findings show that with increased breakage size, the under-expanded jet evolves more rapidly, pressure inside the VV rises more quickly, and the VV reaches its pressure limit sooner. Moreover, localized temperature spikes exceeding material limits are triggered by the convergence of jet-induced compression waves.

In-vessel Loss of Coolant Accident (In-vessel LOCA) in the vacuum vessel of a tokamak fusion reactor, where water is used as the primary coolant, presents a significant challenge. Coolant jets are expected to impinge on high-temperature plasma-facing components (PFCs), resulting in phase transitions on the PFCs surfaces that accelerate pressurization within the vacuum vessel and threaten the chamber's pressure limits. To enhance the predictive capability for pressurization caused by coolant impingement on high-temperature PFCs, the jet impingement heat transfer model must be evaluated and refined. In this study, data from the Ingress of Coolant Event (ICE) experimental setup conducted by the Japan Atomic Energy Research Institute (JAERI) are employed as a benchmark. Initially, a wall heat transfer model based on convective heat transfer is evaluated, revealing a substantial discrepancy between simulation results and experimental data due to the original model's lack of mechanistic considerations, particularly regarding the hysteresis region of impingement heat transfer. Subsequently, two typical jet impingement heat transfer models, developed through experimental and theoretical methods, are introduced. Evaluations indicate that both models underestimate the heat transfer efficiency during coolant impingement on the wall in a vacuum environment, with the Liu model exhibiting errors of approximately 15 % and 10 % for wall temperature and temperature change rate simulations, respectively. The maximum deviation in pressure prediction within the vacuum vessel exceeds 20 %. A theoretical analysis of Helmholtz instability of thin liquid film layer on high-temperature surface in a vacuum environment is then conducted. This analysis considers the enhanced generation of steam under vacuum conditions, leading to a higher steam jet area percentage. Perturbations in the thin liquid film affecting the high-temperature surface are shown to strengthen heat transfer. Based on these findings, a jet impingement heat transfer model tailored for vacuum environments is developed. Comparisons between simulation results and the ICE experimental data demonstrate that the modified model significantly improves predictive accuracy for wall temperature and its rate of change, reducing the pressure prediction error in the vacuum vessel to <10 %.

The aerosol suspended in containment can be removed by the spray system, mitigating the potential risk of radioactive release during severe accidents. To analyze aerosol removal efficiency due to mechanical and phoretic effects of spray droplets, experiments were conducted under various thermal-hydraulic conditions at the Containment Aerosol and Thermal-Hydraulics (CATH) facility. Initial thermal-hydraulic conditions involved either pure air or gas mixture, with a pressure of 5 bar(a), steam fractions of 40%vol and 70%vol, and a temperature range of 20 degrees C-145 degrees C, simulating the severe accident conditions. The sprays were produced by a hollow cone nozzle with a volume mean diameter (VMD) of 370 mu m at a constant spray flow rate. Results indicate that the removal efficiencies of mechanical effects, thermophoresis, and diffusiophoresis are comparable at a steam fraction of 40%vol, while diffusiophoretic effects increase to three times when steam fraction reaches 70%vol. Additionally, the evaluations of the present aerosol removal models show significant deviations from experiments, with inertial impaction is highly sensitive to increases in particle size, while the phoretic mechanism underestimates the effect of steam condensation.

As a critical phenomenon influencing radioactive release assessment during nuclear severe accidents, aerosol resuspension requires in-depth investigation, as it can affect the accuracy of radioactive release source term assessment. An experimental apparatus has been established and the resuspension of multi-layered sedimentary aerosols have been conducted under turbulent airflow conditions with Reynolds number ranging from 50,000 to 130,000. The experiments indicate that higher friction velocity of turbulent pipe and larger deposited particle size both increase the resuspension rate. Through mechanical fulcrum model analysis, the resuspension characteristics, which encompass the coupling effects of airflow characteristics, particle characteristics and wall characteristics, are revealed by the dimensionless particle diameter dp+ and critical dimensionless particle diameter dp50+. A semi-empirical aerosol resuspension model satisfying the S-Logistic function relationship is obtained and validated with multiple sets of experimental data, and showing good agreement between the model predictions and the experimental results.

The coolability of the debris bed with a simulant of solidified corium is experimentally studied, focusing on the effects of the structure of the axial stratified debris bed on the dryout heat flux (DHF). DHF was obtained for the four structures with different particle sizes for the axial stratified debris bed under top flooding. The experimental results show that the dryout position of the axial stratified debris bed is formed at the stratified interface indicated by the temperature rise, and the DHF of the axial stratified bed is much lower than that of the homogeneous bed packed with the upper small particles. To predict the dryout heat flux of the stratified debris beds, by considering the properties of the mixed area, a one-dimensional dryout heat flux model of the porous medium is derived from a water and vapor momentum equation for porous medium, two-phase permeability modifications, interfacial drag, and the correlation between capillary pressure and liquid saturation and verified with the experimental data. The modified model can give reasonable results under different structures.

41Ar是反应堆和加速器运行释放到环境中的主要放射性核素之一，具有半衰期短的特点。由于β-γ符合法能显著降低本底、提高探测器灵敏度，为了实现41Ar高灵敏度测量，设计了由BC404和CsI(Tl)组成的β-γ符合闪烁体探测器，并建立了以最小可探测活度浓度（Minimum Detectable Activity Concentration，MDC）为优化目标的探测器结构优化方法。首先，运用Geant4对β射线在CsI(Tl)中的泄漏率进行模拟，选择BC404厚度为3 mm，此时泄漏率为0.73%；其次，对不同的气体腔尺寸进行模拟，分析计算β探测效率、γ峰效率、气体腔体积和取样时间对MDC的综合影响，完成探测器的结构优化；最后，分析了本底计数率和测量时间对MDC的影响。当测量时间为200分钟、处理时间为30分钟、本底计数率为5×10-3 cps时，优化后的探测器对41Ar的MDC估算值为1.7 Bq/m3。

Sodium-cooled fast reactor (SFR) is one of the most promising Generation IV reactors. Design and safety evaluation of a SFR requires sufficient knowledge about gas bubble behavior in sodium pool. It is challenging to perform numerical study on bubble dynamics in sodium pool due to the discontinuity caused by the differences of physical properties. In this study, a three-dimensional (3D) Moving Particle Semi-implicit (MPS) method is presented featuring high-order accurate scheme and Riemann-based stable multiphase model. Numerical discontinuity caused by large density, large viscosity ratio and surface tension can be handled by the proposed method. To verify the proposed method, two cases from the bubble diagram of Grace (1973) is simulated and rising velocities and shapes of rising gas bubbles are compared, both are agreeing very well. Then co-axial coalescence of gas bubbles are calculated and compared with existing experimental and numerical results. Finally bubble rising behavior in sodium is analyzed using the developed 3D MPS method.

The small modular pressurized water reactors adopt a safety relief system to provide overpressure protection for the reactor coolant system. During this process, the mixed gas is discharged into the suppression water tank, causing direct contact condensation. The fraction of the non-condensable gas has a significant impact on the heat transfer. A visual experimental device was established to conduct the experiments of direct contact condensation within the mixed gas mass flux in the range of 177 similar to 425 kg/m(2)s and the non-condensable gas mass fraction in the range of 0 similar to 20% by using type I multi-holes sparger. The jet plume shapes for the pure steam condensation and mixed gas condensation are investigated in the condensation oscillation zone and jet stable condensation zone. The heat transfer area of pure steam and mixed gas condensation is obtained based on the jet plume shape at the interface. An empirical correlation of the heat transfer coefficient of direct contact condensation with a function of a dimensionless steam mass flux and a driving potential for the condensation process and steam mass fraction is developed and verified with different experimental data. The results show that the empirical correlation of the heat transfer coefficient applies to pure steam and mixed gas, and the error of the model is within +/- 50%.

During the later stages of the postulated Core Disruptive Accidents (CDAs) of a sodium-cooled fast reactor, the porous debris bed is expected to form in the lower plenum region. The geometries of the debris beds are significant for the coolability and the risk of neutronic subcriticality. In this paper, the sedimentation and formation behavior at different factors of the debris bed is experimentally investigated; the mound dimple diameter and mound dimple depth are chosen as the critical quantitative parameters, which can express the flatness of the debris bed intuitively. The results show that the particles formed a concave debris bed at normal temperature. The general tendency of developing a relatively large dimple area is observed with larger nozzle size, lower particle diameter, lower particle density, and higher water level. The prediction model of the mound dimple volume of the debris bed is established, and the error is less than 30%. When the particles are in the film boiling regime, the vapor film reduces the friction between the particle and fluid, thereby increasing the impact momentum and driving the debris bed in a flat direction.

This numerical simulation study has been performed to investigate the dynamic characteristics, oscillation characteristics, and sound pressure of direct contact condensation at low steam mass flux. The volume of fluid model, Ffowcs Williams and Hawkings model, and condensation model are used to simulate the acoustic signal, which is produced during the steam injection process. The simulation results are verified by the experimental results to keep the applicability and accuracy of the model. The results show that the acoustic signal is related to the steam flow patterns. The vibration of the bubble volume, caused by the separation of bubbles, is the important factor for the generation of the acoustic signal. When the subcooled temperature is large, the chugging regime appears during the injection process. The sound pressure amplitude of the steam injection process increases significantly, with the increase in subcooled temperature. In the low-frequency region, which is below 10 Hz, the larger the subcooled temperature increases, the greater the sound pressure amplitude becomes.